JPH0810270B2 - How to carry out spent fuel to the reprocessing facility - Google Patents
How to carry out spent fuel to the reprocessing facilityInfo
- Publication number
- JPH0810270B2 JPH0810270B2 JP5873987A JP5873987A JPH0810270B2 JP H0810270 B2 JPH0810270 B2 JP H0810270B2 JP 5873987 A JP5873987 A JP 5873987A JP 5873987 A JP5873987 A JP 5873987A JP H0810270 B2 JPH0810270 B2 JP H0810270B2
- Authority
- JP
- Japan
- Prior art keywords
- spent fuel
- fuel
- facility
- reprocessing facility
- reprocessing
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired - Lifetime
Links
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims description 44
- 238000012958 reprocessing Methods 0.000 title claims description 37
- 238000000034 method Methods 0.000 claims description 12
- 239000000446 fuel Substances 0.000 description 40
- 238000002844 melting Methods 0.000 description 10
- 230000008018 melting Effects 0.000 description 10
- 238000004090 dissolution Methods 0.000 description 9
- 238000005259 measurement Methods 0.000 description 8
- 238000007726 management method Methods 0.000 description 7
- 238000010008 shearing Methods 0.000 description 7
- 238000012546 transfer Methods 0.000 description 6
- 238000002485 combustion reaction Methods 0.000 description 4
- 238000004886 process control Methods 0.000 description 4
- 239000012634 fragment Substances 0.000 description 3
- 230000002265 prevention Effects 0.000 description 3
- 238000013461 design Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000000712 assembly Effects 0.000 description 1
- 238000000429 assembly Methods 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 238000009499 grossing Methods 0.000 description 1
- 238000010309 melting process Methods 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Description
【発明の詳細な説明】 ［産業上の利用分野］ 本発明は、使用済み燃料集合体の再処理施設に付帯す
る使用済み燃料貯蔵施設に関し、特に該使用済み燃料貯
蔵施設から燃料を再処理施設へ搬出する方法に関するも
のである。Description: TECHNICAL FIELD The present invention relates to a spent fuel storage facility incidental to a spent fuel assembly reprocessing facility, and particularly to a fuel reprocessing facility from the spent fuel storage facility. It is about the method of carrying out to.
［従来の技術］ 原子力発電プラントの使用済み燃料ピットから取り出
される使用済み燃料集合体は、再処理施設に付帯する使
用済み燃料貯蔵池もしくは貯蔵施設への搬入の段階でそ
の燃焼度測定が行なわれてから、前記使用済み燃料貯蔵
施設に貯蔵され、そしてこの貯蔵された使用済み燃料集
合体は、臨界防止及びプロセス管理の観点から、再処理
施設への搬入前に、各種測定原理に基づき再びその燃焼
度測定が行なわれ、再処理の前処理工程、特に溶解槽で
の臨界安全性、作業性の確認を行った上で、使用済み燃
料貯蔵施設より搬出される。より具体的に述べれば、作
業者は、搬出すべき燃料の選択に際して、上述した燃焼
度測定に基づいて、搬出しようとする燃料が、再処理施
設の溶解槽における臨界防止のため核分裂性核種の制限
濃度を満たしうるような燃焼度の燃料かどうか、且つ前
ステツプで装荷した燃料の燃焼度を参考にして、プロセ
スの濃度管理として過度に核分裂性核種の濃度を高くし
ないような燃焼度の燃料かどうかの判断を行っている。[Prior Art] A spent fuel assembly taken out of a spent fuel pit of a nuclear power plant is subjected to burnup measurement at a stage of transportation to a spent fuel storage tank or a storage facility incidental to a reprocessing facility. Then, the spent fuel assembly is stored in the spent fuel storage facility, and the stored spent fuel assembly is again subjected to the measurement based on various measurement principles before being introduced into the reprocessing facility from the viewpoint of criticality prevention and process control. After the burnup is measured and the criticality safety and workability in the pretreatment process of reprocessing, especially in the melting tank are confirmed, it is carried out from the spent fuel storage facility. More specifically, when selecting the fuel to be carried out, the worker determines that the fuel to be carried out is a fissionable nuclide in order to prevent criticality in the melting tank of the reprocessing facility based on the burnup measurement described above. A fuel with a burnup that can satisfy the limit concentration and a burnup that does not excessively increase the concentration of fissile nuclides as the concentration control of the process by referring to the burnup of the fuel loaded in the previous step. I'm making a decision on whether or not.
［発明が解決しようとする問題点］ 従って、際処理施設への搬入前に再び燃焼度の測定を
行い、この測定に基づいて種々の作業を行わねばならな
いので、そのために、長い時間を要するだけでなく、再
処理作業の円滑化が阻害されていた。[Problems to be Solved by the Invention] Therefore, it is necessary to measure the burnup again before carrying it into the reprocessing facility and perform various operations based on this measurement. Therefore, it takes a long time. Not only that, the smoothing of the reprocessing work was hindered.
本発明は、再処理施設へ搬入する際の燃焼度測定にか
かる時間をなくし作業の円滑化を図ると共に、使用済み
燃料貯蔵施設から搬出する段階で再処理施設の臨界安全
管理、プロセス管理の作業能力を高め、再処理施設のプ
ラント稼動率を向上させる、再処理施設への使用済み燃
料搬出方法を提供することを目的とするものである。INDUSTRIAL APPLICABILITY The present invention aims to smooth the work by eliminating the time required to measure the burnup at the time of carrying into the reprocessing facility, and at the stage of carrying out from the spent fuel storage facility, criticality safety management and process management work of the reprocessing facility It is an object of the present invention to provide a method for carrying out spent fuel to a reprocessing facility, which enhances the capacity and improves the plant operating rate of the reprocessing facility.
［問題点を解決するための手段及び作用］ この目的から、本発明による使用済み燃料搬出方法に
おいては、再処理施設における使用済み燃料の核種の濃
度を測定し、その測定値を制御ユニットの演算部に入力
し、該演算部において、前記測定値と前記再処理施設に
おける既知の核種濃度基準値とから、前記制御ユニット
のメモリー部に記憶された使用済み燃料貯蔵施設内の使
用済み燃料の核種の軸方向重量分布に基づいて、前記使
用済み燃料貯蔵施設から前記再処理施設へ搬出される使
用済み燃料を決定し、該使用済み燃料を再処理施設へ搬
出することを特徴としている。核種の軸方向重量分布
は、炉心での燃焼管理及び燃焼により生じた核種の計量
管理のために炉心管理コードに記録されており、これを
制御ユニットのメモリー部に入力しておくことにより、
信頼性の高いこのデータを利用して、再び燃焼度を測定
することなく、再処理施設へ搬出する使用済み燃料を決
定することができる。[Means and Actions for Solving Problems] To this end, in the spent fuel carry-out method according to the present invention, the concentration of the nuclide of the spent fuel in the reprocessing facility is measured, and the measured value is calculated by the control unit. Nuclide of the spent fuel in the spent fuel storage facility stored in the memory unit of the control unit from the measured value and the known nuclide concentration reference value in the reprocessing facility in the arithmetic unit. It is characterized in that the spent fuel to be carried out from the spent fuel storage facility to the reprocessing facility is determined based on the axial weight distribution of, and the spent fuel is delivered to the reprocessing facility. The axial weight distribution of nuclides is recorded in the core control code for combustion control in the core and measurement control of nuclides generated by combustion, and by inputting this in the memory section of the control unit,
This reliable data can be used to determine the spent fuel to be shipped to the reprocessing facility without having to measure burnup again.
［実施例］ 次に、本発明の好適な実施例を添付図面に関連して詳
細に説明する。[Embodiment] Next, a preferred embodiment of the present invention will be described in detail with reference to the accompanying drawings.
第１図は本発明の使用済み燃料搬出方法を実施するシ
ステムの概要を示すもので、同システムには、使用済み
燃料貯蔵池もしくは貯蔵施設Ａに関連して、通常使用さ
れる燃料移送装置１及びその制御ユニット２が設けられ
ており、一般的にはコンピュータである制御ユニット２
は、メモリー部2a、入力部2b、演算部2c及び制御部2dを
有する。この使用済み燃料貯蔵施設Ａは再処理施設Ｂに
付設されている。再処理施設Ｂは周知のように、本発明
に従って燃料移送装置１により使用済み燃料貯蔵施設Ａ
から移送されてきた使用済み燃料（図示せず）を所定の
大きさの小片に剪断する剪断機３と、該剪断機３によつ
て剪断された小片を受け入れる溶解槽４とを備える。FIG. 1 shows an outline of a system for carrying out the spent fuel unloading method of the present invention. In the system, a commonly used fuel transfer device 1 is associated with a spent fuel storage pond or a storage facility A. And its control unit 2 are provided, which is typically a computer
Has a memory unit 2a, an input unit 2b, a calculation unit 2c, and a control unit 2d. The spent fuel storage facility A is attached to the reprocessing facility B. As is well known, the reprocessing facility B is a spent fuel storage facility A with the fuel transfer device 1 according to the present invention.
A shearing machine 3 for shearing the spent fuel (not shown) transferred from the device into small pieces of a predetermined size, and a melting tank 4 for receiving the small pieces sheared by the shearing machine 3.
溶解槽４においては溶解工程が実施され、その後様々
な処理工程が続くが、これ等の工程は本発明の要旨では
ないのでその説明を省略することができる。The melting step is carried out in the melting tank 4, and various processing steps are subsequently performed, but since these steps are not the gist of the present invention, the description thereof can be omitted.
溶解槽４には、その内部の燃料溶解液中の核種（Ｕ、
Pu等）の濃度を測定するための周知の測定器4aが設けら
れており、該測定器4aは制御ユニット２の演算部2cに作
動上接続されていて、測定器4aからの核種濃度測定値が
制御ユニット２の演算部2cに入力されるようになってい
る。精度の向上を図るために核種濃度の測定は適宜行な
われ、その都度新しい情報として入れ換えることができ
る。In the dissolution tank 4, nuclides (U,
A well-known measuring device 4a for measuring the concentration of Pu, etc.) is provided, and the measuring device 4a is operatively connected to the operation unit 2c of the control unit 2 and the nuclide concentration measurement value from the measuring device 4a is provided. Is input to the calculation unit 2c of the control unit 2. In order to improve the accuracy, the nuclide concentration is measured as appropriate and can be replaced with new information each time.
入力部2bに入力されるデータは、炉心での燃焼管理及
び燃焼により生じた核種の計量管理を目的とする炉心管
理コードの出力として原子力発電所より或は設計メーカ
ーより磁気テープ等の媒体として得られるもので、この
データには、使用済み燃料集合体についての燃料番号
（イ）、軸方向の燃焼度分布（ロ）、核種の軸方向重量
分布（ハ）及び平均燃焼度（ニ）が含まれる。また、使
用済み燃料貯蔵施設Ａ内の燃料配置位置（ホ）も入力部
2bに入力されており、これ等のデータ（イ）〜（ホ）は
メモリー部2aに記憶されていて、必要に応じて随時呼び
出すことができる。燃料番号（イ）及び燃料配置位置
（ホ）に関するデータは、排出すべき燃料集合体が決ま
ったら、このデータに基づいて、燃料移送装置１を使用
済み燃料貯蔵施設Ａの該当する燃料集合体の位置へ移動
させ、速やかな燃料の搬出を可能とするために用いられ
る。The data input to the input section 2b is obtained from a nuclear power plant or a design maker as a medium such as a magnetic tape as an output of a core control code for the purpose of combustion control in the core and measurement control of nuclides generated by combustion. This data includes the fuel number (a), axial burnup distribution (b), nuclide axial weight distribution (c) and average burnup (d) for the spent fuel assemblies. Be done. In addition, the fuel arrangement position (e) in the spent fuel storage facility A is also input
The data (a) to (e), which have been input to 2b, are stored in the memory section 2a and can be recalled as needed. The data on the fuel number (a) and the fuel arrangement position (e) is determined based on this data when the fuel assembly to be discharged is determined, and the fuel transfer device 1 is used for the corresponding fuel assembly of the spent fuel storage facility A. It is used to move to the position and enable the quick delivery of fuel.
また、明細書の冒頭に述べたように、従来、搬出燃料
の選択に際しては、搬出しようとする燃料が、再処理施
設の溶解槽における臨界防止のため核分裂性核種の制限
濃度を満たしうるような燃焼度の燃料がどうか、且つ前
ステップで装荷した燃料の燃焼度を参考にして、プロセ
スの濃度管理として過度に核分裂性核種の濃度を高くし
ないような燃焼度の燃料かどうかの判断を行わねばなら
ないが、本発明において制御の対象となるパラメータ
は、臨界管理、プロセス管理の対象となる例えば溶解槽
４中の核分裂性核種の濃度であり、これと、炉心管理コ
ードにより得られた核種の軸方向分布（ハ）を利用する
ことにより、次のステップで再処理施設へ搬送する燃料
を溶解槽４中の核分裂性核種の制限濃度や目標濃度、即
ち基準値に照らし合わせて決定する方法が提供される。Further, as described at the beginning of the specification, conventionally, when selecting an export fuel, it is necessary that the fuel to be exported can satisfy the limiting concentration of fissile nuclides to prevent criticality in the dissolution tank of the reprocessing facility. Whether or not the fuel has a burnup and the burnup of the fuel loaded in the previous step should be used as a reference to determine whether the fuel has a burnup that does not excessively increase the concentration of fissile nuclides as a process concentration control. However, the parameter to be controlled in the present invention is, for example, the concentration of the fissile nuclide in the melting tank 4, which is the target of criticality control and process control, and the axis of the nuclide obtained by the core control code. By using the directional distribution (C), the fuel to be transported to the reprocessing facility in the next step is compared with the limiting concentration or target concentration of the fissile nuclide in the dissolution tank 4, that is, the reference value. Method of determining Te is provided.
さて、炉心管理コードは炉心設計の妥当性を確認し、
炉心を安全に管理するためのものであり、上述したデー
タはこの管理コードに基づいているので、その信頼性は
十分高い。従って、これ等のデータを利用すると共に、
再処理施設Ｂ内の溶解槽４中の核分裂性核種の濃度に基
づいて、再処理施設Ｂへ搬出される燃料を決定すれば、
溶解槽４中の核分裂性核種の濃度は、貯蔵施設Ａより搬
出され、剪断機３に装荷され、溶解槽４へ移送される燃
料、即ち剪断小片に含まれる核分裂性核種の重量に依存
するので、臨界防止やプロセス管理を高い信頼性で実施
しうることが分かる。Now, the core management code confirms the validity of the core design,
It is for the safe management of the core, and the above-mentioned data is based on this management code, so its reliability is sufficiently high. Therefore, while using these data,
If the fuel delivered to the reprocessing facility B is determined based on the concentration of the fissile nuclide in the dissolution tank 4 in the reprocessing facility B,
Since the concentration of the fissile nuclide in the dissolution tank 4 depends on the weight of the fissile nuclide contained in the fuel, that is, the fuel transferred from the storage facility A, loaded into the shearing machine 3, and transferred to the dissolution tank 4, It can be seen that criticality prevention and process control can be performed with high reliability.
今、測定器4aにより求められ演算部2cに入力された現
在の溶解槽４内の核種の濃度をＮ′、臨界防止上の核種
の濃度の制限値或はプロセス管理上の目標値、即ち基準
値をNL、次のステップで溶解槽４に移送される剪断小片
群の核種の濃度をＮとすると、Ｎ′＋Ｎ＜NLであること
が溶解槽４の臨界安全性に必要である。従って、Ｎ＜NL
−Ｎ′としてＮを求めることができ、メモリー部2aに記
憶された情報よりこのＮに相当する剪断予定燃料を有す
る燃料集合体を選択する。例えば、第２図を参照して、
ある使用済み燃料の上部1/4が剪断されると仮定した場
合、メモリー部2aに記憶された炉心管理コード出力の核
種の軸方向重量分布（ハ）より、剪断予定燃料m1におけ
る核分裂性核種の重量N11、N12、・・・N1Nが求まり、
軸方向燃焼度分布（ロ）より、剪断すべき燃料m1の燃焼
度 が求まり、同様にして、m2、m3、m4についても が得られるので、燃料集合体の平均燃焼度 も次式に従って得られる。Now, the present concentration of the nuclide in the dissolution tank 4 obtained by the measuring device 4a and input to the computing unit 2c is N ', a limit value of the concentration of the nuclide for criticality prevention, or a target value for process control, that is, a reference. When the value is N L and the concentration of the nuclide of the sheared fragment group transferred to the dissolution tank 4 in the next step is N, N ′ + N <N L is required for the critical safety of the dissolution tank 4. Therefore, N <N L
N can be obtained as -N ', and the fuel assembly having the scheduled fuel corresponding to N is selected from the information stored in the memory unit 2a. For example, referring to FIG.
Assuming that the upper 1/4 of a certain spent fuel is sheared, from the axial weight distribution (c) of the nuclide of the core control code output stored in the memory unit 2a, the fissile nuclide in the scheduled fuel m 1 is The weights N 11 , N 12 , ... N 1N are calculated,
Burnup of fuel m 1 to be sheared from axial burnup distribution (b) And similarly, for m 2 , m 3 and m 4 Is obtained, the average burnup of the fuel assembly Is also obtained according to the following equation.
これを燃料の決定の指標とすることもできる。ここで、
剪断予定燃料m1よりなる仮定上の剪断小片群中の核種の
濃度Ｎは、N11＋N12＋・・・N1N＝Ｎから求めることが
できる。このようにして、演算部2cにおいて、再処理施
設Ｂへ搬出される燃料を決定したら、メモリー部2aに記
憶された当該燃料の燃料番号（イ）及び燃料配置位置
（ホ）に基づいて、燃料移送装置１を当該燃料の位置に
移動し再処理施設Ｂへ搬出することができる。従って、
この燃料集合体を剪断し剪断小片を溶解槽中に移送して
も、溶解槽の臨界安全性や溶解プロセスには影響がな
く、使用済み燃料貯蔵施設Ａから再処理施設Ｂの剪断機
３への燃料集合体の移送と該剪断機３から溶解槽４への
燃料剪断片の供給とを一貫して安定的に行うことができ
る。 This can also be used as an index for determining the fuel. here,
The concentration N of the nuclide in the hypothetical shear fragment group consisting of the fuel m 1 to be sheared can be obtained from N 11 + N 12 + ... N 1N = N. In this way, when the calculation unit 2c determines the fuel to be carried out to the reprocessing facility B, based on the fuel number (a) and the fuel arrangement position (e) of the fuel stored in the memory unit 2a, The transfer device 1 can be moved to the position of the fuel and taken out to the reprocessing facility B. Therefore,
Even if the fuel assembly is sheared and the sheared pieces are transferred into the melting tank, the critical safety of the melting tank and the melting process are not affected, and the spent fuel storage facility A is transferred to the shearing machine 3 of the reprocessing facility B. It is possible to consistently and stably transfer the fuel assembly and supply the fuel shear fragments from the shearing machine 3 to the melting tank 4.
［発明の効果］ 以上のように、本発明によれば、燃焼度を測定するこ
となく、炉心管理コードのデータを利用して、再処理施
設における使用済み燃料の核種の濃度から、際処理施設
へ搬出する使用済み燃料を決定するので、再処理施設へ
搬入する際の燃焼度測定にかかる時間をなくし作業の円
滑化を図ると共に、使用済み燃料貯蔵施設から搬出する
段階で再処理施設の臨界安定管理、プロセス管理の作業
能力を高め、再処理施設のプラント稼動率を向上させる
ことができる。[Effects of the Invention] As described above, according to the present invention, the core treatment code data is used without measuring the burnup, and the concentration of the nuclide of the spent fuel in the reprocessing facility is used to determine the pretreatment facility. Since the spent fuel to be carried out to the reprocessing facility is determined, the time taken to measure the burnup at the time of importing to the reprocessing facility is eliminated to facilitate the work, and the criticality of the reprocessing facility at the stage of carrying out from the spent fuel storage facility The work capacity of stable management and process management can be improved, and the plant operation rate of the reprocessing facility can be improved.
第１図は、本発明の使用済み燃料搬出方法を実施するシ
ステムの概要を示すブロック図、第２図は、剪断される
使用済み燃料を示す説明図である。 Ａ……使用済み燃料貯蔵施設 Ｂ……再処理施設、１……燃料移送装置 ２……制御ユニット、３……剪断機 ４……溶解槽、2a……メモリー部 2b……入力部、2c……演算部FIG. 1 is a block diagram showing the outline of a system for carrying out the spent fuel carry-out method of the present invention, and FIG. 2 is an explanatory diagram showing the spent fuel to be sheared. A ... spent fuel storage facility B ... reprocessing facility, 1 ... fuel transfer device 2 ... control unit, 3 ... shearing machine 4 ... melting tank, 2a ... memory part 2b ... input part, 2c ...... Calculator
Claims (1)
濃度を測定し、その測定値を制御ユニットの演算部に入
力し、該演算部において、前記測定値と前記再処理施設
における既知の核種濃度基準値とから、前記制御ユニッ
トのメモリー部に記憶された使用済み燃料貯蔵施設内の
使用済み燃料の核種の軸方向重量分布に基づいて、前記
使用済み燃料貯蔵施設から前記再処理施設へ搬出される
使用済み燃料を決定し、該使用済み燃料を再処理施設へ
搬出する使用済み燃料搬出方法。1. A nuclide concentration of a spent fuel in a reprocessing facility is measured, and the measured value is input to a calculation unit of a control unit, and in the calculation unit, the measured value and a known nuclide in the reprocessing facility are measured. Based on the concentration reference value and the axial weight distribution of the nuclides of the spent fuel in the spent fuel storage facility stored in the memory section of the control unit, it is carried out from the spent fuel storage facility to the reprocessing facility. A method for carrying out spent fuel, which determines the spent fuel to be used and carries out the spent fuel to a reprocessing facility.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5873987A JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5873987A JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS63225194A JPS63225194A (en) | 1988-09-20 |
| JPH0810270B2 true JPH0810270B2 (en) | 1996-01-31 |
Family
ID=13092886
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP5873987A Expired - Lifetime JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0810270B2 (en) |
Families Citing this family (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP4643066B2 (en) * | 2001-07-19 | 2011-03-02 | 株式会社東芝 | Reactor fuel reprocessing method, processing sequence determination method, fuel processing planning apparatus and program |
| JP4940114B2 (en) * | 2007-11-30 | 2012-05-30 | 株式会社東芝 | Criticality safety management method for continuous dissolution tank in reprocessing facility |
| JP5558164B2 (en) * | 2010-03-30 | 2014-07-23 | 株式会社東芝 | Fuel treatment planning method, fuel treatment planning system, and fuel treatment planning program |
| JP5422688B2 (en) * | 2012-02-24 | 2014-02-19 | 株式会社東芝 | Criticality safety management method for continuous dissolution tank in reprocessing facility |
-
1987
- 1987-03-16 JP JP5873987A patent/JPH0810270B2/en not_active Expired - Lifetime
Non-Patent Citations (2)
| Title |
|---|
| 原子力工業、３３〔１１〕（１９８７）日刊工業新聞社、Ｐ．１４−２１ |
| 日本原子力学会誌、２８〔８〕（１９８６）（社）日本原子力学会、Ｐ．７３６−７４５ |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS63225194A (en) | 1988-09-20 |
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